National Spherical Tokamak Experiment Research Articles

Density limits as disruption forecasters for spherical tokamaks

Fusion power output from spherical tokamaks would benefit from increased confined plasma density, but there exists a limit on the density before confinement is lost and the plasma current is disrupted. This density limit has long been characterized by a simple, global Greenwald limit proportional to the plasma current and inversely proportional to the cross sectional area of the plasma. It is shown that in the database of discharges from the National Spherical Tokamak Experiment (NSTX) and Mega Ampere Spherical Tokamak (MAST) spherical tokamaks, the likelihood of disruption does increase above the Greenwald limit, and especially in the plasma current rampdown phase. The physics of the density limit has been recently theoretically explored through local criteria. Several of these are tested using the disruption event characterization and forecasting (DECAFTM) code for their potential effectiveness as disruption warning signals. For a limited set of NSTX discharges, a local island power balance criteria was found to be less reliable, presently, than the Greenwald limit. An empirical critical edge line density and a boundary turbulent transport limit were both tested for MAST-U, which has an electron density profile measurement with high spatial resolution in the outer part of the plasma. Both were found to have similar dependencies on key plasma parameters. In a limited set of MAST-U discharges that appear to disrupt due to rising density at values under the Greenwald limit, crossing of the boundary turbulent transport limit occurred close to the time of disruption. Finally, these limits were evaluated for their potential use in real-time, and it was found that with the necessary real-time inputs and with refinement through further testing, these limits could be implemented in a real-time disruption forecasting system.

On the frequency bifurcations of the MHD startup modes in NSTX

The observed bifurcations of the low frequency (<50 kHz) and low toroidal periodicity (n < 5) magnetohydrodynamic (MHD) activity often present in the initial part of the National Spherical Tokamak Experiment (NSTX) discharges can be explained by the evolution of the radial profile of the safety factor (q=rBϕ/RBθ) crossing multiple rational surfaces in the core. Important performance limiting instability mechanisms in the NSTX spherical tokamak are often linked to low frequency and low-n MHD activity. They are quite common in long-pulse NSTX plasmas. They can be present at the beginning of the plasma current flat-top, at the end of the discharge or during the whole duration, and they have been observed to deleteriously impact performance over a wide range of q95. An interesting feature observed in some NSTX discharges is the presence of a bifurcation in the frequency of the low n modes, as low as n = 1, that have frequencies comparable to the plasma core rotation divided by n. Equilibrium reconstructions constrained by magnetic diagnostics data and motional stark effect pitch angle radial profiles suggest that the observed bifurcations are linked to a fast evolving minimum value of q. 3D non-linear resistive MHD simulations show that these modes are ideal and exist as non-resonant before the correspondent rational surface enters the plasma.

Design and Analysis of High Heat Flux Plasma-Facing Components for NSTX Upgrade

The National Spherical Tokamak Experiment (NSTX) upgrade (NSTX-U) requirements lead to enhanced heat loads on plasma-facing components (PFCs) especially in the divertor regions, where normal heat flux density can reach up to 8 MW/m2. For these high heat flux (HHF) regions, the PFCs were redesigned, to use castellations which relieved the thermal stresses associated with high incident heat flux. Another design feature of HHF PFCs is the absence of front face mounting holes which create localized areas of high temperature and thermal stress concentrations. Optimized shaping of the front face of the HHF PFCs eliminates regions where the front face is perpendicular to the heat flux direction during normal helicity operation and spreads the heat load. A unique mechanism was designed to connect PFC tiles to the NSTX-U center stack casing using locking features accessible from the low heat flux regions. Isotropic graphite was selected as the HHF PFC tile material. Initial tile design parameters were accessed using analytical expressions for pulsed heat flux loading. A working prototype of the locking mechanism was created during the initial stages of the design to prove the concept performance. This article presents an overview of the divertor HHF PFC design and the results of the thermal and structural analyses performed using ANSYS software. The results of the analyses cover normal operating conditions and disruptions which impose electromagnetic (EM) loads from eddy and halo currents. The 3-D, transient, nonlinear analyses took into account the temperature-dependent properties of the materials, friction interfaces between the parts, and variable electric properties of the parts and interfaces. The results confirm that the HHF PFC tiles remain within the allowable limits for the loads defined by the NSTX-U Recovery Project. In addition, tolerance stack up analyses were performed to ensure tile performance in the worst possible assembly configuration. The design process was completed successfully, and the NSTX HHF PFC tiles are currently in the production phase.

High Heat Flux Testing of Castellated Graphite Plasma- Facing Components

As part of the recovery project of the National Spherical Tokamak Experiment–Upgrade (NSTX-U), the divertor plasma-facing components (PFCs) were redesigned to handle significantly higher heat fluxes and longer pulse lengths than NSTX. The design process resulted in a castellated, graphite PFC tile. To verify the thermal performance of this design, dedicated electron beam, high heat flux (HHF) testing was carried out on a de-optimized mock-up PFC target. These tests demonstrated that the tile design is itself robust to large, localized thermal gradients. No mechanical damage to the mock-up was observed during HHF testing, though the actual PFC tile mechanical tie-down was not tested. Rather, when the surface temperature exceeded the sublimation temperature of graphite, carbon blooms from the mock-up tile surface were observed. This resulted in 1 to 2 mm of surface material ablating from the mock-up after repeated, highly localized electron beam exposures.

Convolutional Neural Networks for Heat Flux Model Validation on NSTX-U

To demonstrate the use of embedded thermocouples in new National Spherical Tokamak eXperiment Upgrade (NSTX-U) graphite plasma-facing components (PFCs), a convolutional neural network (CNN) has been trained using the ANSYS simulations to predict the scrape-off layer (SOL) heat flux width, $\lambda _{q}$ , given various machine operational parameters and diagnostic data as inputs. The proof-of-concept CNN was trained on the thermocouple data generated by the approximated NSTX-U heat loads applied to real PFC designs in ANSYS. Once trained, the CNN is capable of high precision reconstruction of parameterized heat flux profiles expected in NSTX-U. In addition, to test the system’s ability to cope with noise and systematic error, pseudonoise was injected into the simulated data. CNN can accurately predict the incident heat flux despite this noise and error.

Development of Real-Time Software for Thomson Scattering Analysis at NSTX-U

The Thomson scattering (TS) diagnostic on the National Spherical Tokamak eXperiment Upgrade (NSTX-U) has been an essential system for many operational campaigns due to its function of measuring plasma electron density and temperature. Constructive feedback to improve the next plasma discharge, however, has been limited because of in-between shots analysis. Plasma control, therefore, desires a diagnostic system that is real-time capable. This contribution presents the development of software that demonstrates the feasibility of a real-time TS diagnostic system for NSTX-U. The developed software is able to evaluate the electron temperature and density within 2.5 ms.The overall system requirement is specified by a 60-Hz timing cycle, which is driven by the TS laser pulse rate. The real-time software processes the peak amplitudes of the detected photons, evaluates the electron temperature and density, and then outputs them to an analog output card that is used to interface with the NSTX-U control. The real-time software is implemented in an object-oriented architecture using C++11. C++11 software components include class, Atomic data types for synchronization, and a Hash data structure. The software application makes use of multiple threads that run concurrently: a thread to acquire the photon peak amplitude and feed a circular buffer, threads to evaluate the electron density and temperatures, and a thread that supplies corresponding output voltages and feeds the output card.In summary, the new real-time TS system has been proven to meet the 60-Hz system requirement. For this reason, the software implementation was deemed successful. In future NSTX-U campaigns, this diagnostic will be a great asset enabling real-time plasma density and temperature control.

Initial studies of plasma facing component surface conditioning in the national spherical tokamak experiment upgrade with the materials analysis particle probe

An innovative PFC diagnostic, the Materials Analysis Particle Probe (MAPP) was used to study the chemistry of Plasma Facing Components (PFCs) in the National Spherical Tokamak Experiment Upgrade (NSTX-U). NSTX-U used boronization as conditioning strategy during the 2015–2016 experimental campaign. Deposition with ∼9.1g of deuterated tri-methyl boron (d-TMB) in a helium glow discharge resulted in coatings with an equivalent thickness of 7.0nm at the lower outer divertor region. MAPP was used to capture for the first time the in-vacuo evolution of the chemical state of ATJ graphite PFCs after boronization and plasma exposure via X-ray Photoelectron Spectroscopy (XPS) on a day-to-day basis. The XPS data shows the formation of B4C on the PFCs surface during boronization. We observed the gradual increase of the B2O3 fraction in the coatings with plasma exposures. In contrast, we measured the formation of non-stoichiometric oxides when the samples were only exposed to residual gases and an argon vent. MAPP data revealed erosion and oxidation of deposited boron coatings over the course of tens of shots (a time resolution improved several orders-of-magnitude compared to post-campaign PFC characterization) that is correlated with the transitory nature of plasma performance enhancement with boron conditioning.

Time-Dependent Simulations of Fast-Wave Heated High-Non-Inductive-Fraction H-Mode Plasmas in the National Spherical Torus Experiment Upgrade

30 MHz fast-wave heating may be an effective tool for non-inductively ramping low-current plasmas to a level suitable for initiating up to 12 MW of neutral beam injection on the National Spherical Tokamak Experiment Upgrade (NSTX-U). Previously on NSTX 30 MHz fast wave heating was shown to efficiently and rapidly heat electrons; at the NSTX maximum axial toroidal magnetic field ( B T (0) ) of 0.55 T, 1.4 MW of 30 MHz heating increased the central electron temperature from 0.2 to 2 keV in 30 ms and generated an H-mode plasma with a non-inductive fraction ( f NI ) ∼ 0.7 at a plasma current ( I p ) of 300 kA. NSTX-U will operate at B T (0) up to 1 T, with up to 4 MW of 30 MHz power ( P rf ). Predictive TRANSP free boundary transport simulations, using the TORIC full wave spectral code to calculate the fast-wave heating and current drive, have been run for NSTX-U I p = 300 kA H-mode plasmas. Favorable scaling of f NI with 30 MHz heating power is predicted, with f NI ≥ 1 for P rf ≥ 2 MW.

Charge exchange produced emission of carbon in the extreme ultraviolet spectral region

We used a time-reolving high-resolution grating spectrometer to study extreme ultraviolet emission from plasmas in the National Spherical Tokamak Experiment (NSTX). The NSTX spectral range from 150-250 Å is typically dominated by emission from M-shell iron lines, L- shell transitions of oxygen, or K-shell lines of lithium. However, we also observed several intense emission lines, which we now attribute to transitions in C V and C VI. Collisional-radiative modeling shows that electron-impact excitation is far too weak to account for the features we observed. Instead, these lines appear to be produced by charge exchange with neutral hydrogen.

NSTX-U Control System Upgrades

The National Spherical Tokamak Experiment (NSTX) is undergoing a wealth of upgrades (NSTX-U). These upgrades, especially including an elongated pulse length, require broad changes to the control system that has served NSTX well. A new fiber serial Front Panel Data Port input and output (I/O) stream will supersede the aging copper parallel version. Driver support for the new I/O and cyber security concerns require updating the operating system from Redhat Enterprise Linux (RHEL) v4 to RedHawk (based on RHEL) v6. While the basic control system continues to use the General Atomics Plasma Control System (GA PCS), the effort to forward port the entire software package to run under 64-bit Linux instead of 32-bit Linux included PCS modifications subsequently shared with GA and other PCS users. Software updates focused on three key areas: (1) code modernization through coding standards (C99/C11), (2) code portability and maintainability through use of the GA PCS code generator, and (3) support of 64-bit platforms. Central to the control system upgrade is the use of a complete real time (RT) Linux platform provided by Concurrent Computer Corporation, consisting of a computer (iHawk), an operating system and drivers (RedHawk), and RT tools (NightStar). Strong vendor support coupled with an extensive RT toolset influenced this decision. The new real-time Linux platform, I/O, and software engineering will foster enhanced capability and performance for NSTX-U plasma control.

Emission lines of iron in the 150–250 Å region on National Spherical Torus Experiment

We measured iron emission from the National Spherical Tokamak Experiment. We focused our attention on several band pass regions of the Solar Dynamics Observatory's Atmospheric Imaging Assembly. We found that all significant iron emission in the 171, 193 and 211 Å band pass regions are accounted for by the CHIANTI atomic database, although some strong emission lines of carbon are present that may complicate interpretation of solar data if not taken into account.

Compact “diode-based” multi-energy soft x-ray diagnostic for NSTX

A novel and compact, diode-based, multi-energy soft x-ray (ME-SXR) diagnostic has been developed for the National Spherical Tokamak Experiment. The new edge ME-SXR system tested on NSTX consists of a set of vertically stacked diode arrays, each viewing the plasma tangentially through independent pinholes and filters providing an overlapping view of the plasma midplane which allows simultaneous SXR measurements with coarse sub-sampling of the x-ray spectrum. Using computed x-ray spectral emission data, combinations of filters can provide fast (>10 kHz) measurements of changes in the electron temperature and density profiles providing a method to "fill-in" the gaps of the multi-point Thomson scattering system.

A dual-band adaptor for infrared imaging

A novel imaging adaptor providing the capability to extend a standard single-band infrared (IR) camera into a two-color or dual-band device has been developed for application to high-speed IR thermography on the National Spherical Tokamak Experiment (NSTX). Temperature measurement with two-band infrared imaging has the advantage of being mostly independent of surface emissivity, which may vary significantly in the liquid lithium divertor installed on NSTX as compared to that of an all-carbon first wall. In order to take advantage of the high-speed capability of the existing IR camera at NSTX (1.6-6.2 kHz frame rate), a commercial visible-range optical splitter was extensively modified to operate in the medium wavelength and long wavelength IR. This two-band IR adapter utilizes a dichroic beamsplitter, which reflects 4-6 μm wavelengths and transmits 7-10 μm wavelength radiation, each with >95% efficiency and projects each IR channel image side-by-side on the camera's detector. Cutoff filters are used in each IR channel, and ZnSe imaging optics and mirrors optimized for broadband IR use are incorporated into the design. In-situ and ex-situ temperature calibration and preliminary data of the NSTX divertor during plasma discharges are presented, with contrasting results for dual-band vs. single-band IR operation.

Experimental study of parametric dependence of electron-scale turbulence in a spherical tokamak

Electron-scale turbulence is predicted to drive anomalous electron thermal transport. However, experimental study of its relation with transport is still in its early stage. On the National Spherical Tokamak Experiment (NSTX), electron-scale density fluctuations are studied with a novel tangential microwave scattering system with high radial resolution of ±2 cm. Here, we report a study of parametric dependence of electron-scale turbulence in NSTX H-mode plasmas. The dependence on density gradient is studied through the observation of a large density gradient variation in the core induced by an edge localized mode (ELM) event, where we found the first clear experimental evidence of density gradient stabilization of electron-gyro scale turbulence in a fusion plasma. This observation, coupled with linear gyro-kinetic calculations, leads to the identification of the observed instability as toroidal electron temperature gradient (ETG) modes. It is observed that longer wavelength ETG modes, k⊥ρs≲10 (ρs is the ion gyroradius at electron temperature and k⊥ is the wavenumber perpendicular to local equilibrium magnetic field), are most stabilized by density gradient, and the stabilization is accompanied by about a factor of two decrease in electron thermal diffusivity. Comparisons with nonlinear ETG gyrokinetic simulations show ETG turbulence may be able to explain the experimental electron heat flux observed before the ELM event. The collisionality dependence of electron-scale turbulence is also studied by systematically varying plasma current and toroidal field, so that electron gyroradius (ρe), electron beta (βe), and safety factor (q95) are kept approximately constant. More than a factor of two change in electron collisionality, νe*, was achieved, and we found that the spectral power of electron-scale turbulence appears to increase as νe* is decreased in this collisonality scan. However, both linear and nonlinear simulations show no or weak dependence with the electron-ion collision frequency, νe/i. Instead, other equilibrium parameters (safety factor, electron density gradient, for example) affect ETG linear growth rate and electron thermal transport more than νe/i does. Furthermore, electron heat flux predicted by the simulations is found to have an order-of-magnitude spatial variation in the experimental measurement region and is also found to be much smaller than experimental levels except at one radial location we evaluated. The predicted electron heat flux is shown to be strongly anti-correlated with density gradient, which varies for a factor of three in the measurement region, which is in agreement with the density gradient dependence study reported in this paper.

Simulation of High-Harmonic Fast-Wave Heating on the National Spherical Tokamak Experiment

Images associated with radio-frequency heating of low-confinement mode plasmas in the National Spherical Tokamak Experiment, as calculated by computer simulation, are presented. The AORSA code has been extended to simulate the whole antenna-to-plasma heating system by including both the kinetic physics of the well-confined core plasma and a poorly confined scrape-off plasma and vacuum vessel structure. The images presented show the 3-D electric wave field amplitude for various antenna phasings. Visualization of the simulation results in 3-D makes clear that -30° phasing excites kilo-volt per meter coaxial standing modes in the scrape-off plasma and shows magnetic-field-aligned whispering-gallery type modes localized to the plasma edge.

Predicting High Harmonic Ion Cyclotron Heating Efficiency in Tokamak Plasmas

Observations of improved radio frequency (rf) heating efficiency in ITER relevant high-confinement (H-)mode plasmas on the National Spherical Tokamak Experiment are investigated by whole-device linear simulation. The steady-state rf electric field is calculated for various antenna spectra and the results examined for characteristics that correlate with observations of improved or reduced rf heating efficiency. We find that launching toroidal wave numbers that give fast-wave propagation in the scrape-off plasma excites large amplitude (∼kV m(-1)) coaxial standing modes between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggests that these modes are a probable cause of degraded heating efficiency.

Universal asymptotics of poloidal spectra of magnetic perturbations of saddle coils in tokamaks

Universal and generic features of poloidal spectra of external magnetic perturbations created by a set of saddle coils in tokamak plasmas are studied in a vacuum approximation. It is found that the poloidal mode spectra with high-accuracy are described by a linear combination of three universal asymptotical formulas which depend only on the safety factor of the equilibrium plasma and the geometry of perturbation coils. The validity of these formulas is confirmed by numerical calculations of the mode spectra in the DIII-D plasma [G. L. Jackson et al., Europhys. Conf. Abstr. 27A, P–4 (2007)] and a spherical tokamak, National Spherical Tokamak Experiment [D. A. Gates, et al., Nucl. Fusion 49, 104016 (2009)], plasmas perturbed by external control coils.

Intermittency in the scrape-off layer of the National Spherical Torus Experiment during H-mode confinement

A gas puff imaging diagnostic is used in the National Spherical Tokamak Experiment [M. Ono, et al., Nucl. Fusion 40 (2000) 557] to study the edge turbulence and intermittency present during H-mode discharges. In the case of low power Ohmic H-modes the suppression of turbulence/blobs is maintained through the duration of the (short lived) H-modes. Similar quiescent edges are seen during the early stages of H-modes created with the use of neutral beam injection. Nevertheless, as time progresses following the L–H transition, turbulence and blobs re-appear although at a lower level than that typically seen during L-mode confinement. It is also seen that the time-averaged SOL emission profile broadens, as the power loss across the separatrix increases. These broad profiles are characterized by a large level of fluctuations and intermittent events.

EUV spectroscopy on NSTX

Spectroscopy in the extreme ultraviolet region is important to magnetic fusion research as well as to astrophysics. We report on XEUS (X-ray and Extreme Ultraviolet Spectrometer) and LoWEUS (Long-Wavelength and Extreme Ultraviolet Spectrometer), which operate in the 5–400 Å region on the NSTX (National Spherical Tokamak Experiment) tokamak. The instruments are being used to survey impurities, both for intrinsic elements present in the plasma and for metal impurities resulting from damage to various components. In addition, we have used XEUS and LoWEUS to investigate density-dependent and temperature-dependent emission lines for diagnostic use.

Physics design requirements for the National Spherical Torus Experiment liquid lithium divertor

Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15–25% n e decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, n e/ n GW ∼ 1), to enable n e scan capability (×2) in the H-mode, to test the ability to operate at significantly lower density (e.g., n e/ n GW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m 2) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.